Steam generator flow by-pass system

ABSTRACT

A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger.

TECHNICAL FIELD

The invention relates to a system for removing decay heat from a nuclearreactor.

BACKGROUND

In nuclear reactors designed with passive operating systems, the laws ofphysics are employed to ensure that safe operation of the nuclearreactor is maintained during normal operation or even in an emergencycondition without operator intervention or supervision, at least forsome predefined period of time. A nuclear reactor 5 includes a reactorcore 6 surrounded by a reactor vessel 2. Water 10 in the reactor vessel2 surrounds the reactor core 6. The reactor core 6 is further located ina shroud 122 which surround the reactor core 6 about its sides. When thewater 10 is heated by the reactor core 6 as a result of fission events,the water 10 is directed from the shroud 122 and out of a riser 124.This results in further water 10 being drawn into and heated by thereactor core 6 which draws yet more water 10 into the shroud 122. Thewater 10 that emerges from the riser 124 is cooled down and directedtowards the annulus 123 and then returns to the bottom of the reactorvessel 2 through natural circulation. Pressurized steam 11 is producedin the reactor vessel 2 as the water 10 is heated.

A heat exchanger 35 circulates feedwater and steam in a secondarycooling system 30 in order to generate electricity with a turbine 32 andgenerator 34. The feedwater passes through the heat exchanger 35 andbecomes super heated steam. The secondary cooling system 30 includes acondenser 36 and feedwater pump 38. The steam and feedwater in thesecondary cooling system 30 are isolated from the water 10 in thereactor vessel 2, such that they are not allowed to mix or come intodirect contact with each other.

The reactor vessel 2 is surrounded by a containment vessel 4. Thecontainment vessel 4 is designed so that water or steam from the reactorvessel 2 is not allowed to escape into the surrounding environment. Asteam valve 8 is provided to vent steam 11 from the reactor vessel 2into an upper half 14 of the containment vessel 4. A submerged blowdownvalve 18 is provided to release the water 10 into suppression pool 12containing sub-cooled water.

During a loss of feedwater flow, the nuclear reactor 5 is designed torespond by scramming the reactor core 6, flooding the containment vessel4 or depressurizing the reactor vessel 2. The latter two of theseresponses result in the nuclear reactor 5 being shut down and unable togenerate electricity for an extended period of time. Furthermore, duringa loss of coolant condition where coolant is expelled from the reactorvessel 2, a flow of coolant through the reactor core 6 is reduced. Thisincreases the time needed to bring the reactor core temperatures down toa desired level.

The present invention addresses these and other problems.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 illustrates a nuclear power system.

FIG. 2 illustrates a power module assembly comprising an internally drycontainment vessel.

FIG. 3 illustrates the power module assembly of FIG. 2 during anemergency operation.

FIG. 4 illustrates an embodiment of a power module comprising a steamgenerator flow by-pass system during an emergency operation.

FIG. 5 illustrates an embodiment of a power module comprising a steamgenerator flow by-pass system during normal operating conditions.

FIG. 6A illustrates an embodiment of a steam generator flow by-passsystem during normal operating conditions.

FIG. 6B illustrates an embodiment of the steam generator flow by-passsystem of FIG. 6A during a power-down operation.

FIG. 7 illustrates an embodiment of a steam generator flow by-passsystem comprising a through-passage.

FIG. 8 illustrates an embodiment of a steam generator flow by-passsystem comprising a valve.

FIG. 9 illustrates an embodiment of a steam generator flow by-passsystem comprising one or more baffles.

FIG. 10 illustrates an embodiment of a steam generator flow by-passsystem comprising a temperature activated passage.

FIG. 11 illustrates an embodiment of a steam generator flow by-passsystem comprising a ball check valve.

FIG. 12 illustrates an embodiment of a steam generator flow by-passsystem actuated by control rods.

FIG. 13 illustrates an alternative embodiment of a steam generator flowby-pass system actuated by control rods

FIG. 14 illustrates a novel method of cooling a reactor core using asteam generator flow by-pass system.

SUMMARY OF THE INVENTION

A power module assembly is disclosed as comprising a reactor housing, areactor core located in a lower portion of the reactor housing, and aheat exchanger proximately located about an upper portion of the reactorhousing. The primary coolant flows out of the reactor housing via theupper portion, and the primary coolant flows into the reactor housingvia the lower portion. The power module assembly further comprises apassageway provided in the reactor housing intermediate the lowerportion and the upper portion, wherein the passageway is configured toprovide an auxiliary flow of primary coolant to the reactor core toaugment the flow of the primary coolant out of the upper portion of thereactor housing and into the lower portion.

A nuclear reactor module is disclosed as comprising a reactor vessel anda reactor housing mounted inside the reactor vessel, wherein the reactorhousing comprises a shroud and a riser located above the shroud. A heatexchanger is proximately located about the riser, and a reactor core islocated in the shroud. The nuclear reactor module further comprises asteam generator by-pass system configured to provide an auxiliary flowpath of primary coolant to the reactor core to augment a primary flowpath of the primary coolant out of the riser and into the shroud,wherein the auxiliary flow path of primary coolant exits the reactorhousing without passing by the heat exchanger.

A method of cooling a nuclear reactor is disclosed. A primary coolant iscirculated through a reactor housing comprising an upper riser and alower shroud. A primary flow path of the primary coolant passes by aheat exchanger proximately located about the riser, and the primarycoolant enters the lower shroud. A loss of coolant accident (LOCA) or adepressurization event is detected, and a fluid level of the primarycoolant is decreased below the top of the riser. The primary flow pathof primary coolant exits the riser as steam. An auxiliary flow path ofprimary coolant is circulated through an auxiliary passageway providedin the reactor housing, wherein the auxiliary flow path of primarycoolant exits the reactor housing without passing by the heat exchanger.The primary coolant from the auxiliary flow path is combined with theprimary coolant from the primary flow path that enters the lower shroud.

The invention will become more readily apparent from the followingdetailed description of a preferred embodiment of the invention whichproceeds with reference to the accompanying drawings.

DESCRIPTION OF EXAMPLE EMBODIMENTS

Various embodiments disclosed or referred to herein may be operatedconsistent, or in conjunction, with features found in co-pending U.S.application Ser. No. 11/941,024 which is herein incorporated byreference in its entirety.

FIG. 2 illustrates a power module assembly 50 comprising an internallydry containment vessel 54. The containment vessel 54 is cylindrical inshape, and has spherical, domed, or ellipsoidal upper and lower ends.The entire power module assembly 50 may be submerged in a pool of water16 which serves as an effective heat sink. The pool of water 16 and thecontainment vessel 54 may further be located below ground 9 in a reactorbay 7. The containment vessel 54 may be welded or otherwise sealed tothe environment, such that liquids and gas do not escape from, or enter,the power module assembly 50. The containment vessel 54 may be supportedat any external surface.

In one embodiment, the containment vessel 54 is suspended in the pool ofwater 16 by one or more mounting connections 80. A reactor vessel 52 islocated or mounted inside the containment vessel 54. An inner surface ofthe reactor vessel 52 may be exposed to a wet environment including acoolant 100 or liquid, such as water, and an outer surface may beexposed to a dry environment such as air. The reactor vessel 52 may bemade of stainless steel or carbon steel, may include cladding, and maybe supported within the containment vessel 54.

The power module assembly 50 may be sized so that it can be transportedon a rail car. For example, the containment vessel 54 may be constructedto be approximately 4.3 meters in diameter and 17.7 meters in height(length). Refueling of the reactor core 6 may be performed bytransporting the entire power module assembly 50 by rail car oroverseas, for example, and replacing it with a new or refurbished powermodule assembly which has a fresh supply of fuel rods.

The containment vessel 54 encapsulates and, in some conditions, coolsthe reactor core 6. It is relatively small, has a high strength and maybe capable of withstanding six or seven times the pressure ofconventional containment designs in part due to its smaller overalldimensions. Given a break in the primary cooling system of the powermodule assembly 50 no fission products are released into theenvironment. Decay heat may be removed from the power module assembly 50under emergency conditions.

The reactor core 6 is illustrated as being submerged or immersed in aprimary coolant 100, such as water. The reactor vessel 52 houses thecoolant 100 and the reactor core 6. A reactor housing 20 comprises ashroud 22 in a lower portion and a riser 24 in an upper portion of thereactor housing 20. The shroud 22 surrounds the reactor core 6 about itssides and serves to direct the coolant 100 (shown as coolant flow 65,67) up through the riser 24 located in the upper half of the reactorvessel 52 as a result of natural circulation of the coolant 100. In oneembodiment, the reactor vessel 52 is approximately 2.7 meters indiameter and includes an overall height (length) of 13.7 meters. Thereactor vessel 52 may include a predominately cylindrical shape withellipsoidal, domed or spherical upper and lower ends. The reactor vessel52 is normally at operating pressure and temperature. The containmentvessel 54 is internally dry and may operate at atmospheric pressure withwall temperatures at or near the temperature of the pool of water 16.

The containment vessel 54 substantially surrounds the reactor vessel 52and may provide a dry, voided, or gaseous environment identified ascontainment region 44. Containment region 44 may comprise an amount ofair or other fill gas such as Argonne or other nobel gas. Thecontainment vessel 54 includes an inner surface 55 or inner wall whichis adjacent to the containment region 44. The containment region 44 mayinclude a gas or gases instead of or in addition to air. In oneembodiment, the containment region 44 is maintained at or belowatmospheric pressure, for example as a partial vacuum. Gas or gasses inthe containment vessel may be removed such that the reactor vessel 52 islocated in a complete or partial vacuum in the containment region 44.

During normal operation, thermal energy from the fission events in thereactor core 6 causes the coolant 100 to heat. As the coolant 100 heatsup, it becomes less dense and tends to rise up through the riser 24. Asthe coolant 100 temperature reduces, it becomes relatively denser thanthe heated coolant and is circulated around the outside of the annulus23, down to the bottom of the reactor vessel 52 and up through theshroud 22 to once again be heated by the reactor core 6. This naturalcirculation causes the coolant 100 (shown as coolant flow 65) to cyclethrough the heat exchanger 35, transferring heat to a secondary coolant,such as the secondary cooling system 30 of FIG. 1 to generateelectricity.

FIG. 3 illustrates the power module assembly 50 of FIG. 2 during anemergency operation. The emergency operation may include a response toan overheating of the reactor core 6, or an over-pressurization event ofthe reactor vessel 52, for example. During some emergency operations,the reactor vessel 6 may be configured to release the coolant 100 intothe containment region 44 of the otherwise dry containment vessel 54. Adecay heat of the reactor core 6 may be removed through condensation ofthe coolant 100 on the inner surface 55 of the containment vessel 54.Whereas the containment vessel 54 may be immersed in a pool of water 16,the inner surface 55 of the containment vessel 54 may be completely dryprior to the emergency operation or over-pressurization event.

A flow limiter 58 or steam vent may be mounted on the reactor vessel 52for venting the coolant 100 into the containment vessel 54 during theemergency operation. The coolant 100 may be released into thecontainment vessel 54 as vapor 41, such as steam. The flow limiter 58may be connected or mounted directly to an outer wall of the reactorvessel 52, without any intervening structures such as piping orconnections. The condensation of the vapor 41 may reduce pressure in thecontainment vessel 54 at approximately the same rate that the ventedvapor 41 adds pressure to the containment vessel 54.

Coolant 100 that is released as vapor 41 into the containment vessel 54condenses on the inner surface 55 of the containment vessel 54 as aliquid. The condensation of the vapor 41 causes the pressure in thecontainment vessel 54 to decrease, as the vapor 41 is transformed intothe liquid coolant 100. A sufficient amount of heat may be removed fromthe power module assembly 50 through the condensation of the vapor 41 onthe inner surface 55 of the containment vessel to manage the removal ofdecay heat from the reactor core 6. In one embodiment, there is norelease of the liquid coolant 100 from the reactor vessel 52 even duringan emergency operation. The condensed coolant 100 descends to the bottomof the containment vessel 54 and collects as a pool of liquid. As morevapor 41 condenses on the inner surface 55, the level of the coolant 100in the bottom of the containment vessel 54 gradually rises. Heat storedin the vapor 41 is transferred through the walls of the containmentvessel 54 into the pool of water 16 that acts as an ultimate heat sink.Heat stored in the coolant 100 located at the bottom of the containmentvessel 54 is transferred through liquid convection and conduction heattransfer on the inner surface 55.

Heat removed from the steam or vapor 41 may be transferred to therelatively cold inner surface 55 through condensation on the insidewalls of the cold containment vessel 54 and by natural convection fromthe hot coolant to the inner surface 55. Heat may be transferred to thepool of water 16 by conduction through the containment vessel walls andthrough natural convection on an outside surface of the containmentvessel 54. The coolant 100 remains confined within the power moduleassembly 50 after the reactor core 6 becomes over-heated and during theemergency operation. The heat transferred to the pool of water 16 mayprovide adequate passive decay heat removal for three or more dayswithout any operator intervention.

The containment vessel 54 may be designed to withstand the maximumpressure that would result given an instantaneous release of thehigh-pressure fluid from the reactor vessel 52 into the containmentvessel 54. The pressure inside the containment vessel 54 may be designedto approximately equilibrate with the pressure inside the reactor vessel52, reducing break flow caused by the pressure difference and resultingin coolant level 100A in the reactor vessel 52 and coolant level 100B inthe containment vessel 54 as shown in FIG. 3. The coolant level 100B isshown elevated with respect to the coolant level 100A due to an amountof hydrostatic driving force required for flow through the lower valves57 back into the reactor vessel 52. Differences in coolant levels 100Aand 100B may also exist due to a pressure difference in the reactorvessel 52 relative to the containment vessel 54 due to the pressure dropof the steam flow valve 58. FIG. 3 shows that the coolant levels 100Aand 100B may equilibrate as a result of a hydrostatic head that isgenerated by in imbalance of the coolant levels. Coolant level 100A inthe reactor vessel 52 remains above the top of the reactor core 6,keeping the reactor core 6 covered with coolant 100 at all times. Thecoolant level 100A is maintained by steam or vapor being emitted fromthe riser 24 (shown as coolant flow 42) which condenses on the innersurface 55 of the reactor vessel 52 before collecting at the bottom ofthe reactor vessel 52 to be re-circulated through the reactor core 6.

A flow valve 57 may be provided to allow the coolant 100 to flow fromthe containment vessel 54 back into the reactor vessel 52 once anappropriate or predetermined condition of the coolant levels 100A, 100Bis achieved. Coolant 100 that is allowed to reenter the reactor vessel52 through the flow valve 57 replenishes the coolant 100 that was ventedas vapor 41 through the flow limiter 58. The flow of coolant 100 throughthe flow valve 57 may be achieved through the natural circulation of thepassive system due to the different coolant densities and coolant levelsthat result from temperature differences and valve coolant flow in inthe vessels 52, 54.

Whereas a complete or perfect vacuum may be commercially or technicallyimpractical to achieve or maintain, a partial vacuum may be created inthe containment vessel 54. Any reference to a vacuum herein is thereforeunderstood to be either a partial or complete vacuum. In one embodiment,the containment region 44 is maintained at a vacuum pressure thatsignificantly reduces convective and conductive heat transfer throughthe containment gases. By substantially removing gases from thecontainment region 44, for example by maintaining a vacuum within thecontainment vessel 54, an initial rate as well as subsequent rates ofcondensation of vapor 41 on the inner surface 55 are increased.Increasing the rate of condensation increases the rate of heat transferthrough the containment vessel 54.

In the event of a loss of the vacuum in the containment region 44, theintroduced vapor or liquid provide a further passive safety coolingmechanism to transfer heat between the reactor vessel 52 and thecontainment vessel 54 through natural convection. For example, byreducing or eliminating the thermal insulation, for example as providedby a vacuum, a more effective heat transfer from the reactor vessel 52can be made during an emergency operation due to the condensed liquidcoolant 100 which pools at the bottom of the containment vessel 54. Heatis transferred from the reactor vessel 52 through the liquid coolant 100to the containment vessel 54.

FIG. 4 illustrates an embodiment of a power module assembly 40comprising a steam generator flow by-pass system 45 during an emergencyoperation, such as a loss of coolant accident (LOCA) or anover-pressurization event. Whereas the power module assembly 40 isdescribed with reference to embodiments illustrated in FIGS. 2-3, itshould be understood that many or all of the features could be appliedto the nuclear power system described with respect to FIG. 1 as well asconventional power systems.

A reactor housing 20 is mounted inside the reactor vessel 52, whereinthe reactor housing 20 comprises the shroud 22 and the riser 24 locatedabove the shroud 20. The heat exchanger 35 is proximately located aboutthe riser 24. The reactor core 6 is located in the shroud 22. The riser24 is shown illustrated as being attached to the reactor vessel 52 by anupper attaching member 41, whereas the shroud is shown illustrated asbeing attached to the reactor vessel 52 by a lower attaching member 43.

The steam generator flow by-pass system 45 is configured to provide anauxiliary flow 48 of primary coolant to the reactor core 6 to augment aflow of the primary coolant 100 out of the riser 24 and into the shroud22. The auxiliary flow 48 of primary coolant exits the reactor housing20 without passing by the heat exchanger 35. The steam generator flowby-pass system 45 may provide a hydraulic connection through one or morecomponents of the reactor housing 20. In one embodiment, the steamgenerator flow by-pass system 45 provides a hydraulic connection throughthe annulus (ref. 123 FIG. 1) located intermediate the riser 24 and theshroud 22.

The coolant flow 42 out of the upper portion (e.g. riser 24) of thereactor housing 20 comprises steam, wherein the auxiliary flow 48 ofprimary coolant comprises a mixture of two-phase coolant, such asboiling water. Coolant flow 42 exiting the riser 24 may comprise lesscoolant 100 by mass flow rate as compared to the coolant flow 67 (FIG.2) during normal operations (e.g. full power operation). Auxiliary flow48 may therefore serve to make up some of the lost flow rate, such thatthe coolant flow 46 entering the shroud 22 is augmented to at or nearthe same flow rate as coolant flow 65 in FIG. 2 during normal operation.

In contrast to the coolant level 100N being above the outlet or top ofthe riser 24 shown in FIG. 2 during normal operating conditions, in theembodiment illustrated by FIG. 4 the coolant level 100A is shown belowthe top of the riser 24 during the emergency operation. Whereas thereactor housing 20 is shown completely submerged in primary coolant 100in FIG. 2, the reactor housing 20 is only partially submerged in thecoolant 100 as illustrated in FIG. 4. The level of the primary coolant100 remains above the passageway 45 during normal operation, as well asduring an off-normal operation, shut-down or emergency operation, whensteam generator by-pass occurs.

During normal operating conditions, the coolant flow 65 may be comprisedof predominantly or exclusively single phase coolant, for example in apressurized water reactor design (PWR). Accordingly, a flow of singlephase coolant circulates through the reactor core 6 as coolant flow 65and out the riser 24 as coolant flow 67 (see FIGS. 2 and 5). Thisprovides for single-phase convection heat transfer at the surface of thefuel cladding in the reactor core 6.

When a LOCA occurs and the coolant level 100A drops below the top of theriser 24, as illustrated in FIG. 4, the flow of single phase coolant maybe interrupted. When pressure or temperature variations provide forconditions where the saturation conditions are surpassed, phase-changeheat transfer may occur. Two-phase coolant may develop as it passesthrough the reactor core 6 which may then exit the reactor housing 20via coolant flow 42 as steam which condenses on the inside wall of thereactor vessel 52. By including the auxiliary flow 48 through the steamgenerator flow by-pass system 45, convective heat transfer is providedto the reactor core 6, in addition to the heat transfer that occursthrough steam generation.

The level of coolant 100C within the riser 24 during the LOCA, may dropdown to a level that is approximately equal to that of the coolant level100A on the outside (downcomer) of the reactor housing 20 when the powermodule achieves a steady state condition. A steady state condition mayoccur when the coolant flow 46 entering the shroud 22 is equal to thecombined flow rate of the coolant flow exiting the riser 24 and theauxiliary flow 48 exiting the steam generator flow by-pass system 45.The steam generator flow by-pass system 45 is located above the reactorcore 6 to optimize coolant flow through the fuel rods.

In one embodiment, the steam generator flow by-pass system 45 comprisesa passageway provided in the reactor housing 20 intermediate the lowerportion (e.g. shroud 22) and the upper portion (e.g. riser 24) of thehousing 20, wherein the passageway is configured to provide theauxiliary flow 48 of primary coolant to the reactor core 6 whichaugments the flow of the primary coolant 100 out of the upper portion ofthe reactor housing 20 and into the lower portion. The auxiliary flow 48of primary coolant accordingly bypasses the heat exchanger 35, locatedproximately about the upper portion of the reactor housing 20.

The passageway 45 may be closed during a full power operation of thepower module assembly 40, whereas during an emergency operatingprocedure, the passageway 45 is configured to open. Similarly, thepassageway 45 may be configured to open during a shut-down, orpower-down operation, including a LOCA or over-pressurization event. Inone embodiment, the passageway remains open during all modes ofoperation, whereas the auxiliary flow 48 is substantially minimized orreduced to zero during normal operations of the power module assembly40.

FIG. 5 illustrates an embodiment of a power module comprising a steamgenerator flow by-pass system 59 during normal operating conditions. Thesteam generator flow by-pass system 59 comprises an opening orpassageway through the reactor housing 120. For example, the passagewaymay be located between or through a lower end 60 of the riser 24 and anupper end 62 of the shroud 22. The coolant flow 65 passes through thereactor core 6 located in the shroud 22 before exiting the riser 24 ascoolant flow 67. During normal operations, little or none of the coolantflow 65 escapes through the steam generator flow by-pass system 59. Byprohibiting or reducing a flow rate through the steam generator flowby-pass system 59, a maximum flow of coolant passes by the heatexchanger 35 to remove heat from the reactor core 6. Accordingly, themass flow rate of coolant flow 65 is approximately equal to that ofcoolant flow 67.

FIG. 6A illustrates an embodiment of a steam generator flow by-passsystem 69 during normal operating conditions, such as when a powermodule is operating at full power. During normal operation, the powermodule generates an operating temperature that is typically higher thana temperature associated with reactor start-up, reactor shut-down, orother operating conditions. Different temperatures may be generated atdifferent locations within the coolant 100 as a result of interactionwith the heat exchanger 35 (FIG. 4). At normal operating temperatures,coolant flow 65 and 67 behave substantially as described with respect toFIGS. 2 and 5. Different components of the reactor housing 20 mayundergo different amounts of thermal expansion, as a result of thedifference in operating temperature or as a result of differences inthermal properties of the various components. For example, somecomponents may be made out of different materials, composition, oramount (e.g. thickness), such that one component may expand or retractto a greater degree than another component.

In one embodiment, a direction of expansion or contraction of the shroud22 and the riser 24 are in opposite directions. For example, while theriser 24 expands toward the bottom of the reactor vessel 52 (FIG. 2),the shroud 22 expands toward the top of the reactor vessel 52. Thisrelationship is diagrammatically illustrated by the downward and upwardfacing arrows at the lower end 60 of the riser 24 and the upper end 62of the shroud 22, respectively. Expansion of the components in oppositedirections may be accomplished by attaching the riser 24 to the upperattaching member 41 and by separately attaching the shroud 22 to thelower attaching member 43 (FIG. 4).

A passageway 63 in the upper end 62 of the shroud 22 is shown dislocatedwith a passageway 61 in the lower end 60 of the riser 24. With theshroud 22 and riser 24 in the thermally expanded condition, thedislocated passageways 61, 63 do not line up, such that little or noneof the coolant flow 65 is allowed to pass through the steam generatorflow by-pass system 69.

FIG. 6B illustrates an embodiment of the steam generator flow by-passsystem 69 of FIG. 6A during a power-down operation. The power downoperation may include a reactor shut-down, reactor trip or SCRAM, LOCA,or overpressurization event, for example. During the power-downoperation, temperatures in the reactor vessel 52 (FIG. 2) tend todecrease, which results in a contraction or retraction of variousreactor components. For example, while the riser 24 contracts toward thetop of the reactor vessel 52 (FIG. 2), the shroud 22 retracts toward thebottom of the reactor vessel 52. This relationship is diagrammaticallyillustrated by the upward and downward directed arrows at the lower end60 of the riser 24 and the upper end 62 of the shroud 22, respectively.The riser 24 and the shroud 22 may expand or contract at differentamounts for the same change in temperature, in which case the directionsof expansion and retraction may be relative to each other.

The passageway 63 in the upper end 62 of the shroud 22 is shown alignedwith the passageway 61 in the lower end 60 of the riser 24, allowing anauxiliary flow 48 of coolant to pass through the steam generator flowby-pass system 69. With the shroud 22 and riser 24 in the thermallyretracted condition, the co-located passageways 61, 63 line-up to form athrough-passage, such that the auxiliary flow 48 is combined withcoolant flow 42. In one embodiment, the passageway 61, 63 opens due to achange in temperature within the reactor vessel 52 (FIG. 2), wherein adifference in rate of thermal expansion between the shroud 22 and theriser 24 causes the passageway 61, 63 to open. A flow rate of theauxiliary flow 48 may vary according to the change in temperature, adegree of alignment between the passageways 61, 63, or the number ofpassageways provided in the reactor housing 20. The auxiliary flow 48 ofcoolant exits the reactor housing 20 without passing by or through theheat exchanger 35 (FIG. 4).

FIG. 7 illustrates an embodiment of a steam generator flow by-passsystem 79 comprising a through-passage 70. The through-passage 70 may beformed between the lower end 60 of the riser 24 and the upper end 62 ofthe shroud 22. The lower end 60 and upper end 62 are shown overlappingeach other, such that the auxiliary flow 48 circulates through thethrough-passage 70. FIG. 7 may be understood as representing the flow ofcoolant during a shut-down or power-down operation, in which coolantflow 42 provides a reduced flow rate as compared to coolant flow 67 ofFIG. 5. In FIG. 5, during normal operation of the power module 40,coolant flows 65, 67 may be sufficiently strong such that little or noauxiliary flow escapes from the steam generator flow by-pass system 59.Flow paths through the riser 24 may provide the path of least resistanceduring normal operation.

During a shut-down operation, or LOCA, where coolant flow 42 may bereduced, auxiliary flow 48 may be allowed to exit the through-passage 70through natural convection, as coolant flow 46 exceeds the flow rate ofcoolant flow 42. In one embodiment, the primary coolant exits the steamgenerator flow by-pass system 79 as a result of a decrease in flow rateof the coolant flow 42 of the primary coolant out of the riser 24. Thedecrease in flow rate may correspondingly decrease an amount of eddiesthat otherwise form in the through-passage 70 during normal operatingconditions, allowing the coolant to “boil over” through the steamgenerator flow by-pass system 79.

In the embodiment illustrated in FIG. 7 as well as the other variousembodiments described and illustrated herein, the auxiliary flow 48 ofprimary coolant may exit the reactor housing 20 due to naturalconvection, or natural circulation of the coolant. A two-phase state ofthe coolant may promote auxiliary flow 48 of coolant to pass through thesteam generator flow by-pass system, whereas most or all of the coolantwould otherwise exit out the riser 24 when the coolant is in asingle-phase state (e.g. during normal operating conditions). Passivelycooling the reactor core 6 (FIG. 5) reduces or eliminates the need forproviding moving or mechanical parts, such as motors.

In one embodiment, a distance between the overlapped section of thelower end 60 and upper end 62 increases or decreases with a change intemperature of the power module 40. During a decrease in reactortemperature, forces F1 and F2 may act on the ends 60, 62 of the riser 24and shroud 22 to increase the size of the through-passage 70 and providefor an increase in auxiliary flow 48. Whereas during an increase inreactor temperature, the size of the through-passage 70 may decrease asthe distance between the overlapped section of the ends 60, 62decreases, resulting in the auxiliary flow 48 decreasing or ceasing toflow. A flow rate of the auxiliary coolant 48 may vary with a change inreactor temperature and associated change in size or flow area of thethrough-passage 70.

FIG. 8 illustrates an embodiment of a steam generator flow by-passsystem 89 comprising a valve 80 positioned near the lower end 60 of theriser 24 and the upper end 62 of the shroud 22. Auxiliary flow 48 may beallowed to flow similarly as with regards to the description of FIG. 7,whereas the valve 80 may be provided to limit a direction of the coolantflow 48 in a single direction. In one embodiment, valve 80 is aunidirectional valve that limits the direction of coolant flow 48 fromwithin the reactor housing 20 to outside of the reactor housing 20. Inone embodiment, the valve 80 is always open, and the rate of auxiliaryflow 48 is governed by the flow rate of coolant flow 42, 46 or coolantflow 65, 67 (FIG. 5). In another embodiment, valve 80 is actuated (e.g.opened) upon detection of a shut-down operation or reactor scram, forexample, such that valve 80 is otherwise closed during normal (e.g. fullpower) reactor operation.

FIG. 9 illustrates an embodiment of a steam generator flow by-passsystem 99 comprising one or more baffles 90. The auxiliary flow 48through the baffles 90 may operate or function similarly as describedabove with respect to the embodiments illustrated in FIGS. 4-8. Forexample, during normal operation of the power module 40, little or noauxiliary flow 48 may be allowed to exit through the one or more baffles90. During a shut-down operation, auxiliary flow 48 through the baffles90 may be enabled or increased.

In one embodiment, the one or more baffles 90 rotate about a pivot toopen or close. Baffle 90A illustrates a baffle in a closed position,whereas baffle 90B illustrates a baffle in an open position. The one ormore baffles 90 may open or close depending on the flow rate of thecoolant flow 42, 46, as these flow rates may exert pressure P1, P2 onthe one or more baffles 90. If a flow rate or pressure differentialbetween pressures P1, P2 is great enough, the one or more baffles 90 mayclose, and prohibit a flow of coolant through the steam generator flowby-pass system 99. The steam generator flow by-pass system 99 mayfurther comprise a return mechanism, such as a spring, that returns theone or more baffles 90 to an open position when the flow rate dropsbelow some predetermined threshold. In one embodiment, the steamgenerator flow by-pass system 99 comprises a screen with miniaturelouvers or baffles that allow the passage of boiling coolant, butprohibit or limit the passage of single phase coolant.

FIG. 10 illustrates an embodiment of a steam generator flow by-passsystem 109 comprising a temperature activated passage 100. The passage100 may be configured to open due to a change in temperature within thereactor vessel 52 (FIG. 4). In one embodiment, the steam generator flowby-pass system 109 comprises a bi-metallic cover located over thepassageway, wherein the bi-metallic cover comprises materials havingdifferent thermal expansion rates or properties. In one embodiment, thepassageway is formed between the riser 24 and the shroud 22. A first endof the temperature activated passage 100 may be fixed or otherwiseattached to the reactor housing 20 (FIG. 4). Due to the differentthermal expansion properties, a second end of the temperature activatedpassage 100 may bend away from the reactor housing 20 with a force Fo asa reactor temperature decreases. A passageway through the reactorhousing 20 may therefore be formed which allows the auxiliary flow 48 toexit the steam generator flow by-pass system 109.

As the reactor temperature increases, the temperature activated passage100 may relax, or bend back to cover the passageway (shown by reference100A) and reduce or stop the auxiliary flow 48 from exiting the reactorhousing 20.

FIG. 11 illustrates an embodiment of a steam generator flow by-passsystem 119 comprising a ball check valve 110. The ball check valve 110may move in a bi-direction sense, such that in one position it allowsthe auxiliary flow 48 to pass through the steam generator flow by-passsystem 119, whereas in a second position (e.g. shown as reference 110A)it limits or prohibits the release of auxiliary flow 48 out of thereactor housing 20.

The steam generator flow by-pass system 119 may comprise a return spring115 that urges the ball check valve 110 toward the open, first position.The amount of force exerted by the return spring 115 may exceed theforce due to the coolant flow 48 during a shut-down condition, forexample. During normal operation, a flow rate due to coolant flow 65(FIG. 5) may overcome the force exerted by the return spring 115, andplace the ball check valve 110 in the closed, second position 110A. Inanother embodiment, the weight of the ball in the ball check valveprovides the downward force of the ball check valve 110, replacing theneed for the return spring 115.

In another embodiment, a spring is located near the bottom of the ballcheck valve 110, instead of as shown in FIG. 11. The spring expandsduring normal operation due to an increase in temperature, urging theball check valve 110 toward the closed, second position 110A. The springcontracts during a power down condition due to a decrease intemperature, urging the ball check valve 110 toward the open, firstposition.

FIG. 12 illustrates an embodiment of a steam generator flow by-passsystem 129 actuated by control rods 125A, 125B. The steam generator flowby-pass system 129 may comprise one or more vents or valves 120 attachedto the reactor housing 20. In one embodiment, the steam generator flowby-pass system 129 is attached to the reactor housing 20 intermediatethe shroud 22 and the riser 24.

When the control rods (identified as reference number 125B) are removedfrom the reactor core 6, the steam generator flow by-pass system 129 maybe actuated to be closed, such that little or no auxiliary flow isallowed to exit the reactor housing 20. The steam generator flow by-passsystem 129 may be closed, for example, during normal or full-poweroperation of the power module 40. When the control rods (identified asreference number 125A) are inserted into the reactor core 6, the steamgenerator flow by-pass system 129 may be actuated to be open, such thatthe auxiliary flow is allowed to exit the reactor housing 20. The steamgenerator flow by-pass system 129 may be open, for example, duringshut-down or a power down operation of the power module 40. One or moreswitches or sensors may determine when the control rods 125A, 125B areinserted or removed from the reactor core 6, and send a signal toactuate the steam generator flow by-pass system 129.

FIG. 13 illustrates an alternative embodiment of a steam generator flowby-pass system 139 actuated by control rods 135A, 135B. The steamgenerator flow by-pass system 139 may comprise one or more control rodsdesigned such that when withdrawn (135A) for operation they obstruct theflow path of the by-pass system, and when inserted (135B) during powerdown conditions they provide an open passage to auxiliary coolantby-pass flow 48. The location of the control rods 135A, 135B allow orprevent the auxiliary flow of primary coolant to pass through thehousing 20. In one embodiment, the steam generator flow by-pass system139 is attached to the reactor housing 20 intermediate the shroud 22 andthe riser 24.

One or more switches or sensors may determine when the control rods areinserted (135B) or removed (135A) from the reactor core 6.

FIG. 14 illustrates a novel method of cooling a reactor core using asteam generator flow by-pass system. The method may be understood tooperate with, but not limited by, various embodiments illustrated hereinas FIGS. 1-13.

At operation 140 a primary coolant is circulated through a reactorhousing comprising an upper riser and a lower shroud, wherein a primaryflow path of the primary coolant passes by a heat exchanger proximatelylocated about the riser, and wherein the primary coolant enters thelower shroud.

At operation 150, a loss of coolant accident (LOCA) or adepressurization event is detected. The LOCA or depressurization eventmay indicate a reduced amount of coolant or pressure in the reactorvessel.

At operation 160, a fluid level of the primary coolant is decreasedbelow the top of the riser, wherein the primary coolant exits the riseras steam. In one embodiment, the primary coolant that exits the riser assteam condenses as liquid coolant before being combined with anauxiliary flow path of the primary coolant that is circulated through anauxiliary passageway in the reactor housing.

At operation 170, the auxiliary flow path of the primary coolant iscirculated through the auxiliary passageway provided in the reactorhousing, wherein the auxiliary flow path of the primary coolant exitsthe reactor housing without passing by the heat exchanger. In oneembodiment, the auxiliary flow path of the primary coolant circulatesthrough the auxiliary passageway due to a difference in hydrostaticforces on either side of the passageway.

At operation 180, the primary coolant from the auxiliary flow path iscombined with the primary coolant from the primary flow path that entersthe lower shroud.

In one embodiment, chemical additives soluble in coolant of a nuclearreactor are combined with the primary coolant of a nuclear reactor,modifying the nuclear and chemical characteristics of the coolant. Aloss of primary coolant inventory is detected, and a fluid level of theprimary coolant is decreased such that the nominal flow path isinterrupted or reduced. Production of steam occurs in the core region,and exits the riser as steam. Non volatile additives in the primarycoolant are concentrated in the core, and coolant devoid of thenon-volatile additives collects in regions observing condensation. Theprimary coolant is circulated through a passageway provided in thereactor housing, wherein the coolant devoid of additives is combinedwith the coolant with increased concentration of additives, providingmixing of the coolant streams and mitigating the concentration process.Circulating the auxiliary flow path of the primary coolant through theauxiliary passageway reduces a concentration of non-volatile additivesin the primary coolant within the reactor housing.

Although the embodiments provided herein have primarily described apressurized water reactor, it should be apparent to one skilled in theart that the embodiments may be applied to other types of nuclear powersystems as described or with some obvious modification. For example, theembodiments or variations thereof may also be made operable with aboiling water reactor or more generally to any other integrated passivereactor design.

The rate of release of the coolant into the containment vessel, the rateof condensation of the coolant into a liquid, and the rate of increaseof pressure in the containment vessel, as well as other rates and valuesdescribed herein are provided by way of example only. Other rates andvalues may be determined through experimentation such as by constructionof full scale or scaled models of a nuclear reactor.

Having described and illustrated the principles of the invention in apreferred embodiment thereof, it should be apparent that the inventionmay be modified in arrangement and detail without departing from suchprinciples. We claim all modifications and variation coming within thespirit and scope of the following claims.

The invention claimed is:
 1. A power module assembly comprising: areactor vessel that comprises a substantially sealed enclosure; aprimary fluid coolant enclosed in the sealed enclosure of the reactorvessel; a reactor core located in a lower portion of the reactor vessel,the reactor core comprising a primary fluid coolant outlet near an upperend of the reactor core and a primary fluid coolant inlet near a lowerend of the reactor core; a riser conduit that extends from near the topof the reactor core to an upper portion of the reactor vessel; a heatexchanger located about a portion of the riser conduit in the upperportion of the reactor vessel; a fluid bypass path defined between alower end portion of the riser conduit and the upper end of the reactorcore and substantially enclosed within the sealed enclosure of thereactor vessel between the primary fluid coolant outlet of the reactorcore and the heat exchanger, the fluid bypass path hydraulicallycoupling a fluid coolant path that extends between the primary fluidcoolant inlet of the reactor core and the primary fluid coolant outletof the reactor core with an annulus between the riser conduit and thereactor vessel; and a flow restriction positioned within the fluidbypass path and between the lower end portion of the rise conduit andthe upper end portion of the reactor core.
 2. The power module assemblyaccording to claim 1, wherein during a loss of coolant accident, theflow of primary coolant out of the upper portion of the riser conduitcomprises steam, and wherein a flow of primary coolant through the fluidbypass path comprises a mixture of two-phase coolant.
 3. The powermodule assembly according to claim 1, wherein the fluid bypass path isclosed or reduced during a full power operation of the power moduleassembly.
 4. The power module assembly according to claim 3, wherein thefluid bypass path is configured to open during a shut-down operation. 5.The power module assembly according to claim 4, wherein the shutdownoperation comprises a loss of coolant accident or an over pressurizationevent.
 6. The power module assembly according to claim 1, wherein alevel of the primary coolant is above an outlet of the upper portion ofthe reactor vessel during full power operation, and wherein the level ofprimary coolant is below the outlet during a shut-down operation.
 7. Thepower module assembly according to claim 6, wherein the level of theprimary coolant remains above the fluid bypass path during the shut-downoperation.
 8. The power module assembly according to claim 1, whereinthe flow restriction comprises a unidirectional valve that defines aflow path only from the fluid coolant path that extends between theprimary fluid coolant inlet of the reactor core and the primary fluidcoolant outlet of the reactor core to the annulus.
 9. The power moduleassembly according to claim 1, wherein the flow restriction comprisesone of: an always-open valve; or a modulating valve that comprises anactuator configured to adjust the modulating valve between an openposition and a closed position.
 10. A nuclear reactor module comprising:a reactor vessel that comprises a substantially sealed enclosure; areactor housing mounted inside the reactor vessel, the reactor housingcomprising a shroud and a riser located above the shroud within thesealed enclosure of the reactor vessel; a heat exchanger proximatelylocated about the riser; a reactor core located in the shroud; a fluidbypass path defined between a lower end portion of the riser and anupper end portion of the shroud and substantially enclosed within thesealed enclosure of the reactor vessel between a fluid outlet of theshroud and the heat exchanger, the fluid bypass path hydraulicallycoupling a fluid coolant path that extends between a fluid inlet of thereactor core and the fluid outlet of the shroud with an annulus betweenthe riser and the reactor vessel; and a flow restriction positionedwithin the fluid bypass path and between the lower end portion of theriser and the upper end portion of the shroud.
 11. The nuclear reactormodule according to claim 10, wherein an auxiliary flow of primarycoolant exits the reactor housing due to a difference in hydrostaticforces in the fluid bypass path between the fluid coolant path thatextends between the fluid inlet of the reactor core and the fluid outletof the shroud and the annulus.
 12. The nuclear reactor module accordingto claim 11, wherein the primary coolant exits the reactor housing as aresult of a decrease in rate of the primary flow path of the primarycoolant out of the riser.
 13. The nuclear reactor module according toclaim 10, wherein the fluid bypass path forms a passageway for coolantto exit the reactor housing during a loss of coolant accident or adepressurization event.
 14. The nuclear reactor module according toclaim 10, wherein the shroud comprises a nozzle-shaped member, the fluidoutlet of the shroud being smaller than a fluid inlet of the shroud. 15.The nuclear reactor module according to claim 14, wherein the fluidinlet of the shroud is substantially the same size as a fluid outlet ofthe reactor core.
 16. The nuclear reactor module according to claim 10,wherein the flow restriction comprises a unidirectional valve thatdefines a flow path only from the fluid coolant path that extendsbetween the fluid inlet of the reactor core and the fluid outlet of theshroud to the annulus.
 17. The nuclear reactor module according to claim10, wherein the flow restriction comprises one of: an always-open valve;or a modulating valve that comprises an actuator configured to adjustthe modulating valve between an open position and a closed position. 18.A nuclear reactor module comprising: a reactor vessel that comprises asubstantially sealed enclosure; a reactor housing mounted inside thereactor vessel, the reactor housing comprising a shroud and a riserlocated above the shroud within the sealed enclosure of the reactorvessel; a heat exchanger proximately located about the riser; a reactorcore located in the shroud; and a fluid bypass path defined between alower end portion of the riser and an upper end portion of the shroudand substantially enclosed within the sealed enclosure of the reactorvessel between a fluid outlet of the shroud and the heat exchanger, thefluid bypass path hydraulically coupling a fluid coolant path thatextends between a fluid inlet of the reactor core and the fluid outletof the shroud with an annulus between the riser and the reactor vessel,wherein an auxiliary flow of primary coolant exits the reactor housingdue to a difference in hydrostatic forces in the fluid bypass pathbetween the fluid coolant path that extends between the fluid inlet ofthe reactor core and the fluid outlet of the shroud and the annulus. 19.The nuclear reactor module according to claim 18, wherein the shroudcomprises a nozzle-shaped member, the fluid outlet of the shroud beingsmaller than a fluid inlet of the shroud, and the fluid inlet of theshroud is substantially the same size as a fluid outlet of the reactorcore.
 20. The nuclear reactor module according to claim 18, furthercomprising a flow restriction positioned within the fluid bypass pathand between the lower end portion of the riser and the upper end portionof the shroud, the flow restriction comprising at least one of: aunidirectional valve that defines a flow path only from the fluidcoolant path that extends between the fluid inlet of the reactor coreand the fluid outlet of the shroud to the annulus; an always-open valve;or a modulating valve that comprises an actuator configured to adjustthe modulating valve between an open position and a closed position. 21.A power module assembly comprising: a reactor vessel that comprises asubstantially sealed enclosure; a primary fluid coolant enclosed in thesealed enclosure of the reactor vessel; a reactor core located in alower portion of the reactor vessel, the reactor core comprising aprimary fluid coolant outlet near an upper end of the reactor core and aprimary fluid coolant inlet near a lower end of the reactor core; ariser conduit that extends from near the top of the reactor core to anupper portion of the reactor vessel; a heat exchanger located about aportion of the riser conduit in the upper portion of the reactor vessel;and a fluid bypass path defined between a lower end portion of the riserconduit and the upper end of the reactor core and substantially enclosedwithin the sealed enclosure of the reactor vessel between the primaryfluid coolant outlet of the reactor core and the heat exchanger, thefluid bypass path hydraulically coupling a fluid coolant path thatextends between the primary fluid coolant inlet of the reactor core andthe primary fluid coolant outlet of the reactor core with an annulusbetween the riser conduit and the reactor vessel, wherein during a lossof coolant accident, the flow of primary coolant out of the upperportion of the riser conduit comprises steam, and wherein a flow ofprimary coolant through the fluid bypass path comprises a mixture oftwo-phase coolant.
 22. A power module assembly comprising: a reactorvessel that comprises a substantially sealed enclosure; a primary fluidcoolant enclosed in the sealed enclosure of the reactor vessel; areactor core located in a lower portion of the reactor vessel, thereactor core comprising a primary fluid coolant outlet near an upper endof the reactor core and a primary fluid coolant inlet near a lower endof the reactor core; a riser conduit that extends from near the top ofthe reactor core to an upper portion of the reactor vessel; a heatexchanger located about a portion of the riser conduit in the upperportion of the reactor vessel; and a fluid bypass path defined between alower end portion of the riser conduit and the upper end of the reactorcore and substantially enclosed within the sealed enclosure of thereactor vessel between the primary fluid coolant outlet of the reactorcore and the heat exchanger, the fluid bypass path hydraulicallycoupling a fluid coolant path that extends between the primary fluidcoolant inlet of the reactor core and the primary fluid coolant outletof the reactor core with an annulus between the riser conduit and thereactor vessel, wherein the fluid bypass path is closed or reducedduring a full power operation of the power module assembly.
 23. Thepower module assembly according to claim 22, wherein the fluid bypasspath is configured to open during a shut-down operation, and theshutdown operation comprises a loss of coolant accident or an overpressurization event.
 24. A power module assembly comprising: a reactorvessel that comprises a substantially sealed enclosure; a primary fluidcoolant enclosed in the sealed enclosure of the reactor vessel; areactor core located in a lower portion of the reactor vessel, thereactor core comprising a primary fluid coolant outlet near an upper endof the reactor core and a primary fluid coolant inlet near a lower endof the reactor core; a riser conduit that extends from near the top ofthe reactor core to an upper portion of the reactor vessel; a heatexchanger located about a portion of the riser conduit in the upperportion of the reactor vessel; and a fluid bypass path defined between alower end portion of the riser conduit and the upper end of the reactorcore and substantially enclosed within the sealed enclosure of thereactor vessel between the primary fluid coolant outlet of the reactorcore and the heat exchanger, the fluid bypass path hydraulicallycoupling a fluid coolant path that extends between the primary fluidcoolant inlet of the reactor core and the primary fluid coolant outletof the reactor core with an annulus between the riser conduit and thereactor vessel, wherein a level of the primary coolant is above anoutlet of the upper portion of the reactor vessel during full poweroperation, and wherein the level of primary coolant is below the outletduring a shut-down operation.
 25. The power module assembly according toclaim 24, wherein the level of the primary coolant remains above thefluid bypass path during the shut-down operation.